3.13

Molten Salt Reactor Fuel and Coolant

O. Benesˇ and R. J. M. Konings
European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe, Germany

ß 2012 Elsevier Ltd. All rights reserved.

3.13.1

Introduction

360

3.13.2
3.13.3
3.13.4
3.13.4.1
3.13.4.2
3.13.4.2.1
3.13.4.2.2
3.13.4.2.3
3.13.4.2.4
3.13.4.2.5
3.13.4.2.6
3.13.4.2.7
3.13.4.2.8
3.13.4.2.9
3.13.4.2.10
3.13.4.3

3.13.4.3.1
3.13.4.3.2
3.13.4.3.3
3.13.4.3.4
3.13.4.3.5
3.13.4.3.6
3.13.4.4
3.13.4.4.1
3.13.4.4.2
3.13.4.4.3
3.13.4.4.4
3.13.4.4.5
3.13.4.4.6
3.13.4.5
3.13.4.5.1
3.13.4.5.2
3.13.4.5.3
3.13.4.5.4
3.13.4.5.5
3.13.4.5.6
3.13.4.6
3.13.4.6.1
3.13.4.6.2
3.13.4.6.3
3.13.4.6.4
3.13.4.6.5
3.13.4.6.6
3.13.5
3.13.6


Historical Background
Fuel Concepts of MSR
Properties of the MSR Fuels and Coolants
Structural Aspects of Molten Salts
Phase Diagrams
LiF–BeF2
LiF–PuF3
NaF–PuF3
BeF2–PuF3
BeF2–ThF4
LiF–AnF4
LiF–BeF2–AnF4
LiF–NaF–BeF2–AnF3
NaF–NaBF4
LiF–NaF–KF
Solubility of Actinides in the Fluoride Melt
ThF4 in molten LiF
ThF4 in molten LiF–BeF2
UF4 in molten LiF–ThF4
PuF3 in molten LiF–BeF2
PuF3 in molten LiF–NaF–BeF2
PuF3 in molten LiF–BeF2–ThF4
Density and Viscosity
LiF–BeF2
LiF–AnF4
LiF–BeF2–ThF4
LiF–NaF–BeF2–AnF4
NaF–NaBF4
LiF–NaF–KF
Heat Capacity and Thermal Conductivity

LiF–BeF2
LiF–AnF4
LiF–BeF2–ThF4
LiF–NaF–BeF2–PuF3
NaF–NaBF4
LiF–NaF–KF
Vapor Pressure
LiF–BeF2
LiF–AnF4
LiF–BeF2–ThF4
LiF–NaF–BeF2–AnF3
NaF–NaBF4
LiF–NaF–KF
Role of Oxygen Impurities
Electroanalytical Chemistry

361
362
363
363
365
365
365
366
366
367
367
369
370
371

371
371
371
371
373
373
373
373
374
374
374
374
375
376
376
377
377
377
378
379
379
379
379
379
379
380
380
380
381
381

381
359


360

Molten Salt Reactor Fuel and Coolant

3.13.7
3.13.8
3.13.8.1
3.13.8.2
3.13.8.3
3.13.8.4
3.13.9
3.13.10
References

Radiation Stability of Molten Salts
Fission Product Behavior
Noble Gases
Salt-Soluble Fission Products
Insoluble Fission Products
Iodine
The Effect of Corrosion Reactions on the Fuel Behavior
Summary and Future Work

Abbreviations
AHTR
ARE

CNRS

Advanced high-temperature reactor
Aircraft Reactor Experiment
Centre National de la Recherche
Scientifique
FLIBE
Eutectic mixture of LiF and BeF2
MOSART Molten Salt Actinide Recycler and
Transmuter
MSBR
Molten salt breeder reactor
MS-FR
Molten salt cooled fast reactor
MSFR
Molten salt fast reactor
MSR
Molten salt reactor
MSRE
Molten Salt Reactor Experiment
ORNL
Oak Ridge National Laboratory
PWR
Pressurized water reactor
SFR
Sodium cooled fast reactor
VHTR
Very high-temperature reactor

3.13.1 Introduction

The molten salt reactor (MSR) is one of the six
reactor concepts of the Generation IV initiative,
which is an international collaboration to study the
next generation nuclear power reactors. The fuel
of the MSR is based on the dissolution of the fissile
material (235U, 233U, or 239Pu) in an inorganic liquid
that is pumped at a low pressure through the reactor
vessel and the primary circuit, and thus also serves
as the primary coolant. The heat generated by the
fission process is transferred in a heat exchanger to
a secondary coolant, which is also generally a molten
salt. This intermediate loop is introduced for safety
reasons: to avoid direct contact between the steam
and the fuel. A schematic drawing of the MSR is
shown in Figure 1 as taken from US DOE Roadmap.1
The operating temperature of the MSR is between
800 and 1000 K, the lower limit being determined by
the fusion temperature of the salt and the upper one

382
383
384
384
385
385
385
386
387

by the corrosion rate of the structural material

(see Chapter 5.10, Material Performance in Molten
Salts). Typical inlet and outlet temperatures of some
MSR concepts, which are briefly discussed in Section
3.13.3, are summarized in Table 1. It is worth mentioning that at least a 50 K safety margin must be kept
in all concepts, and hence the melting temperature of
the fuel salt must be at least 50 K lower than the
designed inlet temperature of the reactor.
The fact that the fuel of the MSR is in the liquid
state offers several advantages. The first among them is
the safety of the reactor. As the fuel is in the liquid state
and serves as primary coolant having low vapor pressures (boiling points >1400  C), the total pressure of
the primary circuit is kept very low (p $ 1 bar) compared to, for example, current light water reactors. It
thus avoids the major driving force, the high pressure,
for radioactivity release during accidents. Another
aspect that contributes to the safety of the MSR is
that the reactor possesses a strong negative temperature coefficient, so the chain reaction automatically
slows down when the temperature increases. This is
induced by the thermal expansion of the primary coolant, which pushes the fuel out of the reactor core (the
fuel density decreases). The third characteristic that
increases the safety of the reactor is the possibility of
draining the liquid fuel into emergency dump tanks
in case of an accident. The emergency tanks are
installed under the reactor and are designed in such
way that the fuel remains in a subcritical state.
Another big advantage of the MSR is the possibility of performing a continuous fuel cleanup,
which results in an increase of the fuel burnup.
This chemical cleanup can be done either online
or in batches. The goal of the fuel cleanup is to
separate the fission products from the fuel and transfer them into the nuclear waste, while the cleaned
fuel is sent back into the primary circuit. It is very

important to make this separation because most of
the fission products have a very high neutron capture


Molten Salt Reactor Fuel and Coolant

MSR

Molten salt reactor

361

Control
rods

Coolant salt

Reactor

Electrical
power

Generator

Purified
salt
Turbine

Fuel salt


Pump

Heat
exchanger

Chemical
processing
plant

Recuperator
Heat
exchanger

Compressor
Freeze
plug

Heat
sink

Pump

Precooler

Heat sink

Intercooler
Compressor

Emergency dump tanks


Figure 1 Schematic drawing of the molten salt reactor. Reproduced from US DOE Nuclear Energy Research Advisory
Committee and the Generation IV International Forum, A Technology Roadmap for Generation IV Nuclear Energy Systems,
© Generation IV International Forum.

Table 1
Typical fuel salt inlet and outlet temperatures of
some MSR concepts
MSR concept

Tinlet

Toutlet

References

MSRE
MSBR
MSR FUJI
MSFR
MOSART

908 K
839 K
840 K
903 K
873 K

936 K
977 K

980 K
923 K
988 K

2
3
4
5
6

cross-section and thus slow down the chain reaction.
Because of the online cleanup, a very low amount of
fission products is present in the fuel during the
reactor operation, and thus the heat generation from
their radioactive decay is small and the risk of overheating in the event of loss of cooling is avoided.
Moreover, it is also possible to profit from the neutron economy and design the MSR as a breeder reactor that produces more fuel than it consumes, for
example, using a 232Th/233U cycle.
Furthermore, because of the liquid state of the
MSR fuel, there is no radiation damage to the fuel

(as discussed in Section 3.13.7). Therefore, issues
such as swelling or crack formation that appear in
the case of ceramic fabricated fuels are avoided.

3.13.2 Historical Background
The first proposal for a MSR dates back to the 1940s
when Bettis and Briant proposed it for aircraft propulsion.7 A substantial research program was started
at the Oak Ridge National Laboratory (ORNL) in
the United States to develop this idea, culminating in
the Aircraft Reactor Experiment (ARE) that went

critical for several days in 1954. However, no airplane
with such propulsion has ever been constructed.
For ARE, a mixture of NaF–ZrF4 was used as carrier
of the fissile UF4 for the following reasons8,9:





Wide range of solubility for thorium and uranium
Thermodynamic stability up to high temperatures
No radiolytic decomposition
Low vapor pressure at the operating temperature
of the reactor


362

Molten Salt Reactor Fuel and Coolant

 Compatibility with nickel-based alloys (Ni–Mo–
Cr–Fe) that can be used as structural materials.
In the second half of the 1950s, the molten salt
technology was transferred to the civilian nuclear
program of the United States. At the time, many
reactor concepts were being studied and the interest
in breeder reactors was growing. It was recognized
that the MSR would be ideal for thermal breeding of
uranium from thorium,7 and the Molten Salt Reactor
Experiment (MSRE) was started at ORNL to demonstrate the operability of MSRs. Because of the

breeding aspect, the neutron economy in the reactor
was considered to be of key importance, and 7LiF–
BeF2 (FLIBE), with 5% ZrF4 as oxygen getter, was
selected as fuel carrier because of the very low neutron capture cross-sections of 7Li (sthermal ¼ 0.045
barn) and Be (sthermal ¼ 0.0088 barn). Natural lithium cannot be used as part of the nuclear fuel as
it contains about 7.6% of 6Li (the remaining 92.4%
is 7Li), which has a very high parasitic neutron
capture cross-section (sthermal ¼ 940 barn). Therefore, enrichment of 7Li is required before it can be
used as a fuel matrix. The MSRE was a graphitemoderated reactor of 8 MWth (megawatt thermal)
and operated from 1965 to 1969. Two different fissile
sources were used: initially, 235UF4 was used with
33% enrichment and later, 233UF4 was added to the
carrier salt, making the MSRE the world’s first reactor to be fueled with this fissile material.10 FLIBE was
used as coolant in the secondary circuit. The results
of MSRE, which have been reported in great detail,10
revealed that all the selected materials (fuel, structurals) behaved well and that the equipment behaved
as predicted. In this respect, it was very successful.
After the MSRE, a design for a prototype molten
salt breeder reactor (MSBR) was made by ORNL in
the early 1970s.3 The program was stopped in 1976
in favor of the liquid metal cooled fast reactor7:
although the technology was considered promising,
there were technological problems that had to be
solved. The MSBR design was a 2250 MWth reactor,
optimized to breed 233U from 232Th in a single fluid
system. Online pyrochemical cleanup was planned to
clean the fuel solvent from the neutron-absorbing
fission products. Nevertheless, interruption of reactor
operation was planned every 4 years to replace the
graphite moderator, as experiments had revealed significant swelling of graphite due to radiation damage.

Because of the (semi)continuous online clean up of
the fuel, the addition of zirconium to the fuel was not
necessary, and FLIBE could be used as carrier of the

fertile (ThF4) and fissile elements (UF4). As secondary coolant, a NaF–NaBF4 (8–92 mol%) mixture was
foreseen, particularly because the tritium retention
of this salt is much better than FLIBE.
In the 1990s, there was a renewed interest
in molten salt technology, which originated from
programs that were looking into the possibilities
of transmutation of actinides. When addressing
transmutation of minor actinides, the absence of
complicated fuel and fuel pin fabrication and the
compatibility with pyrochemical processing in the
molten salt fuel cycle were recognized as important
advantages, in comparison with conventional pellet
fuel types. Also, the interest in the use of thorium as a
nuclear fuel kept up the interest in MSRs. As a result,
the MSR is now one of the six reactor concepts
selected for the Generation IV initiative, which is
looking at next generation nuclear reactors. Current
MSR designs, however, move away from thermal
graphite-moderated concepts, and favor nonmoderated concepts that have a fast(er) neutron spectrum.
Fuel selection for the nonmoderated reactor concepts is more flexible, and elements other than 7Li
can be considered. One reason is that the neutron
capture cross-section of the alkali halides and alkaliearth halides is generally lower in the ‘fast’ spectrum
than in the thermal spectrum; also, the neutron economy is not as sensitive in the ‘fast’ spectrum as in
the thermal one. Therefore compounds like NaF,
KF, RbF, or CaF2 can be considered as part of
the fuel matrix. Moreover, there are some ‘fast’

MSR concepts, for example, the REBUS-3700 concept,11 which are based on the chloride matrix (35Cl:
sfast ¼ 0.0011 barn, whereas sthermal ¼ 43.63 barn).

3.13.3 Fuel Concepts of MSR
The fuel in the MSR must fulfill several requirements with respect to its physicochemical properties
(as will be discussed in Section 3.13.4). These
requirements are very well met by the various systems containing alkali metal and alkali-earth fluorides; hence the fluoride systems are the most
recognized candidates for MSR fuels.
In the previous section, the MSBR has been mentioned as a graphite-moderated reactor that is based
on the 7LiF–BeF2–232ThF4–UF4 system.3 232ThF4 is
a fertile material that is used to produce fissile 233UF4
by a neutron capture and two consecutive b-decays
of 233Th and 233Pa. This fuel composition based on
the FLIBE matrix still remains an ideal candidate


Molten Salt Reactor Fuel and Coolant

when the MSR is designed as a thermal breeder
reactor (moderated reactor). In this case, neutron
economy is very critical and only isotopes with very
low neutron capture cross-section in the thermal
spectrum can be part of the fuel matrix. Thus, 7LiF
and BeF2 are the prime compounds for consideration.
One of the current MSR concepts that uses fuel
technology similar to that of the MSBR is the MSR
FUJI concept.4 Originally proposed by Furukawa, it
is a rather small graphite-moderated concept with an
installed thermal capacity of 450 MW.
Nowadays the nonmoderated reactors are attracting interest because they offer the possibility of transmuting the long-lived actinides produced mostly in

light water reactors. The transmutation is most effective in the fast neutron spectrum; however, due to the
presence of the fluorine atom in the fuel, partial
moderation is maintained, and the neutron spectrum
of the MSR is, rather, shifted to the epithermal range.
Nevertheless, at this energy, all the minor actinides
are fissionable, and the fission-to-capture ratio for
these nuclides is still much higher than in the thermal
spectrum.12 Furthermore, the nonmoderated reactor
does not require graphite blocks (moderator in the
thermal MSR) in the reactor core: they are very
susceptible to radiation damage and must be periodically replaced.
At the moment, there are two main directions
for the nonmoderated MSR concepts. The first is
an actinide burner design based on the Russian
MOSART (Molten Salt Actinide Recycler and Transmuter) concept,6 for which the 7LiF–(NaF)–BeF2–
AnF3 system is proposed as a fuel salt. The startup
and feed material scenarios can include plutonium
and minor actinides from pressurized water reactor
(PWR) spent fuel. Depending upon the feed material,
the fuel salt at equilibrium contains 0.7–1.3 mol% of
actinide and lanthanide trifluorides. The second one
is an innovative concept called MSFR (molten salt fast
reactor), which has been developed by Centre
National de la Recherche Scientifique (CNRS) in
France.5,13–16 The fuel in this concept is based on
the 7LiF–232ThF4 matrix, with the addition of actinide fluorides as a fissile material. There are two
initial fissile choices in the MSFR concept: (1) the
233
U-started MSFR and (2) the transuranic-started
MSFR with a mix of 87.5% of Pu (238Pu 2.7%, 239Pu

45.9%, 240Pu 21.5%, 241Pu 10.7%, and 242Pu 6.7%),
6.3% Np, 5.3% of Am, and 0.9% of Cm in the form of
fluorides, corresponding to the transuranic element
composition of a UO2 fuel after one use in a PWR and
5 years of storage.17

363

One of the very recent MSR designs is the
REBUS-3700 concept, which is based on a chloride
salt as a fuel. It is a fast breeder reactor proposed by
Mourogov and Bokov11 and it is based on a 238U/239Pu
cycle, where 238U serves as a fertile material bred
to fissile 239Pu by neutron capture and two consecutive b-decays of 239U and 239Np. Both uranium and
plutonium are present in the form of trichlorides
dissolved in a matrix of liquid NaCl. In general, the
chlorides have higher vapor pressures and lower
thermodynamic stability at high temperatures compared to fluorides, but, on the other hand, their
melting points are lower. Therefore, more fissile
material can be dissolved in the matrix, which is
essential for fast breeder reactor designs. However,
the chlorides can be used only in fast reactors and not
in thermal ones due to the relatively high parasitic
neutron capture cross-section of the chlorine atom,
as already discussed in Section 3.13.2.
A summary of various applications of molten salts
in future nuclear reactor designs is given in Table 2.
As the primary choices for the MSR fuels or coolants
are based on the fluoride systems, the chloride systems are not discussed further.


3.13.4 Properties of the MSR Fuels
and Coolants
In this section, the physicochemical properties of the
primary MSR fuel and coolant choices from Table 2
are discussed, with the emphasis on the melting
behavior, actinide solubility in the fuel matrix, density, viscosity, heat capacity, thermal conductivity, and
vapor pressure. All these quantities are highly relevant for the reactor design calculations and a summary of these properties for typical coolant and fuel
compositions is given in Tables 3 and 4 respectively.
Optimized phase diagrams of the relevant fluoride
systems used as MSR fuels, coolants, or heat transfer
salts are also shown in this section.

3.13.4.1

Structural Aspects of Molten Salts

Molten fluoride salts are essentially ionic liquids in
which cations and anions form a loose network. Some
cations occur in their simplest form, such as Liþ and
Naþ, but some form molecular species like BeF2,
which is a structural analogue to SiO2, known to be
highly associated and forming a network structure
that exhibits a glass transition characteristic. In a


364

Molten Salt Reactor Fuel and Coolant

Table 2


The various applications of molten salts in nuclear reactor concepts

Reactor type

Neutron
spectrum

Application

Primary choice

MSR breeder

Thermal
Fast

Fuel
Fuel
Secondary coolant
Fuel
Primary coolant
Heat transfere
Primary coolant
Intermediate coolantf

7

MSR burner
AHTRa

VHTRb
MS-FRc
SFRd

Fast
Thermal
Thermal
Fast
Fast

LiF–BeF2–AnF4
LiF–AnF4
NaF–NaBF4
LiF–NaF–BeF2–AnF3
7
LiF–BeF2
LiF–NaF–KF
LiCl–NaCl–MgCl2
NaNO3–KNO3
7

Alternative(s)

7

LiF–CaF2–AnF4, NaCl–UCl3–PuCl3
LiF–BeF2, KF–KBF4
LiF–NaF–KF–AnF3, LiF–NaF–RbF–AnF3
LiCl–KCl–MgCl2


a

Advanced high-temperature reactor, graphite-moderated, thermal reactor.
Very high-temperature reactor, graphite-moderated, gas cooled reactor.
c
Molten salt cooled fast reactor, the solid fuel fast reactor with MS as a coolant.
d
Sodium cooled fast reactor.
e
Heat transfer salt is a medium that will be used to deliver heat from the reactor to the hydrogen production plant.
f
To separate sodium and the steam circuits.
b

Table 3

Selected properties of the coolant salts

Property

LiF–BeF2 (0.66–0.34)

NaF–NaBF4 (0.08–0.92)

LiF–NaF–KF (0.465–0.115–0.42)

Melting point (K)
r(kg mÀ3)
(mPa s)
Cp(J KÀ1 gÀ1)

l(W mÀ1 KÀ1)
log10(p(Pa))

728
2146.3–0.4884T (K)
1.81exp(1912.2/T (K))
2.39
1.1
11.914–13003/T (K)

657 Æ 1
2446.3–0.711T (K)
0.0877exp(2240/T (K))
1.506
0.66–2.37 Â 10À4T (K)
11.638–6550.6/T (K)

727
2579.3–0.6240T (K)
0.0248exp(4477/T (K))
1.88
0.36 þ 5.6 Â 10À4T (K)
10.748–10789/T (K)

Table 4

Selected properties of the fuel salts

Property


LiF–ThF4 (0.78–0.22)

LiF–BeF2–ThF4 (0.717–0.16–0.123)

LiF–NaF–BeF2–PuF3
(0.203–0.571–0.212–0.013)

Melting point (K)
r(kg mÀ3)
(mPa s)
Cp(J KÀ1 gÀ1)
l(W mÀ1 KÀ1)
log10(p(Pa))

841
5543.0–1.2500 T (K)
0.365exp(2735/T (K))
1.0
$1.5a
11.902–12 989/T (K)

771
4124.3–0.8690 T (K)
0.062exp(4636/T (K))
1.55
1.5a
11.158–10 790.5/T (K)

775
2759.9–0.5730 T (K)

0.100exp(3724/T (K))
2.15
0.402 þ 0.5 Â 10À3/T (K)
11.6509–12 827/T (K)

Value for T ¼ 1023 K.

a

recent study by Salanne et al.,18 a molecular dynamic
study was performed on the LiF–BeF2 system in order
to understand the structure of the (Li,Be)F2Àx melt.
Figure 2 shows the distribution of various species
observed in the solution as a function of BeF2
composition. At low concentrations of BeF2 in LiF,
the mixture behaves as a well-dissociated ionic melt
À
consisting of Liþ, BeF2À
4 , and F species. As BeF2
concentration increases, the BeF2À
units start to
4
bond together sharing a common FÀ ion, first creat7À
ing Be2 F3À
7 species, followed by Be3 F10 species, and

so forth, resulting in a polymer of several BeF2À
4
units. This polymerization is also a reason why the
viscosity of pure BeF2 is much higher compared to

that of other fluorides discussed in this chapter.
species were also experimentally observed
BeF2À
4
by spectroscopic studies, as reported by Toth and
Gilpatrick.19 Lanthanide fluorides, ThF4 or PuF3
also form molecular species in their liquid form,
but in comparison to BeF2, they do not exhibit polymerization. Dracopolous et al.20,21 investigated
the structure of molten KF–YF3 and KF–LnF3


Molten Salt Reactor Fuel and Coolant

365

100

80
FBeF4260
%F

Be2F734-

Be3F10

40

5-

Be4F13


‘Polymer’
20

0

0

20

40
mol% BeF2

60

80

Figure 2 Percentage of F atoms involved in various species observed in the LiF–BeF2 system as a function of
composition; ‘polymer’ means a cluster with a Be nuclearity >4, whereas FÀ implies that the ion is coordinated only to Liþ.
Reproduced from Salanne, M.; Simon, C.; Turq, P. J. Phys. Chem. B 2007, 111, 4678–4684.

(Ln ¼ La, Ce, Nd, Sm, Dy, Yb) systems using Raman
spectroscopy and found that at x(LnF3) 0.25, LnF3À
6
are the predominant species surrounded by Kþ
cations. At higher concentrations of LnF3, the
lanthanides are forced to share common fluorides
and start to create loose structures of bridged
octahedra. On the basis of these two studies, the
authors concluded that lanthanide melts have similar structural behavior. In case of thorium, a tetravalent ion is the only known species in molten

fluorides. As reported by Barton,22 ThF4 forms
mainly anionic complexes of the general formula
23
À
ThFmÀ
4þm , and the existence of ThF5 is claimed. In
case of uranium, tri- or tetravalent ions are stable in
the molten fluoride salt. It has been demonstrated19
that UF4 dissolves in the fluoride melts, forming
complexes of coordination numbers 7 or 8. It has
been shown that in fluoride-rich systems, the UF4À
8
species predominates, while with the reduction of
fluoride ions, the UF3À
7 species is produced according
3À
À
to UF4À
8 Ð UF7 þ F . Furthermore, the same
authors confirmed that approximately equal amounts
3À
of UF4À
8 and UF7 occur in the LiF–BeF2 melt of
intermediate composition.
3.13.4.2

Phase Diagrams

3.13.4.2.1 LiF–BeF2


The LiF–BeF2 phase diagram has been assessed by
van der Meer et al.24 and more recently by Benesˇ

and Konings,25 the latter version being preferred
as the authors considered not only the equilibrium points measured,26–28 but also the mixing
enthalpies of the (Li,Be)Fx liquid solution
measured by Holm and Kleppa.29 The LiF–BeF2
phase diagram is shown in Figure 3; it is characterized by two eutectic invariant equilibria found at
T ¼ 636 K and xðBeF2 Þ ¼ 0:517, and T ¼ 729 K and
xðBeF2 Þ ¼ 0:328 in the calculation. Two intermediate phases, Li2BeF4 and LiBeF3, are present in
the system as well, the first melting congruently
at T ¼ 729 K, whereas the latter decomposes below
the solidus at T ¼ 557 K. A miscibility gap appears
in the BeF2-rich side, with the monotectic temperature found at T ¼ 772 K, while the critical
temperature was found at Tc ¼ 812 K and
x(BeF2) ¼ 0.826.
3.13.4.2.2 LiF–PuF3

The thermodynamic assessment of the LiF–PuF3
system was made in a study by van der Meer et al.30
and later by Benesˇ and Konings,31 using a different
thermodynamic model based on the equilibrium data
measured by Barton and Strehlow.32 The calculated
phase diagram as obtained from the data of Benesˇ
and Konings is shown in Figure 4, indicating very
good agreement with the experimental data. The
system is characterized by a single eutectic at
T ¼ 1018 K and x(PuF3) ¼ 0.212.



366

Molten Salt Reactor Fuel and Coolant

1300

1100

T (K)

900

700

500

300

0

0.2

0.6

0.4

0.8

1


x (BeF2)
Figure 3 Calculated LiF–BeF2 phase diagram from Benesˇ and Konings25: ◊ experimental data by Roy et al.26; □ data by
Thoma et al.27; and △ data by Romberger et al.28 Reproduced from Benesˇ, O.; Konings, R. J. M. J. Chem. Thermodyn.
2009, 41, 1086–1095.

1800

1600

T ( K)

1400

1200

1000

800
0.0

0.2

0.4

0.6

0.8

1.0


x (PuF3)
Figure 4 The calculated LiF–PuF3 phase diagram based on the thermodynamic data taken from Benesˇ and Konings31:
○ experimental data measured by Barton and Strehlow.32 Reproduced from Benesˇ, O.; Konings, R. J. M. J. Nucl.
Mater. 2008, 377(3), 449–457.

3.13.4.2.3 NaF–PuF3

Similar to the LiF–PuF3 system, the NaF–PuF3
phase diagram has been thermodynamically assessed
in two studies,30,31 both based on the experimental
data measured by Barton et al.33 The phase diagram
is shown in Figure 5 and is characterized by one
eutectic at T ¼ 999 K and x(PuF3) ¼ 0.221 and one

peritectic at T ¼ 1111 K and x(PuF3) ¼ 0.387, where
the NaPuF4 intermediate compound decomposes.
3.13.4.2.4 BeF2–PuF3

To our best knowledge, there are no published experimental data on the BeF2–PuF3 system. Benesˇ and
Konings25 made a thermodynamic assessment of this


Molten Salt Reactor Fuel and Coolant

367

1800

1600


T ( K)

1400

1200

1000

800
0.0

0.2

0.4

0.6

0.8

1.0

x (PuF3)
Figure 5 The calculated NaF–PuF3 phase diagram based on the thermodynamic data taken from Benesˇ and Konings31:
○ experimental data measured by Barton et al.33 Reproduced from Benesˇ, O.; Konings, R. J. M. J. Nucl. Mater. 2008,
377(3), 449–457.

1800

1500


T ( K)

1200

900

600

300
0.0

0.2

0.4

0.6

0.8

1.0

x (PuF3)
Figure 6 The estimated BeF2–PuF3 phase diagram. Reproduced from Benesˇ, O.; Konings, R. J. M. J. Chem.
Thermodyn. 2009, 41, 1086–1095.

system, assuming an ideal behavior of the liquid
phase. The estimated BeF2–PuF3 phase diagram is
shown in Figure 6, consisting of a single eutectic
point at T ¼ 783 K and x(PuF3) ¼ 0.031.


by Thoma et al.34 The calculated phase diagram is
shown in Figure 7. It is a simple eutectic system with
the eutectic at T ¼ 800 K and x(ThF4) ¼ 0.019.

3.13.4.2.5 BeF2–ThF4

The LiF–ThF4 system is a reference salt for the MSFR
concept. The equilibrium diagram of the LiF–ThF4
system was reported by Thoma et al.35 on the

The BeF2–ThF4 system was assessed by van der
Meer et al.24 using the equilibrium data measured

3.13.4.2.6 LiF–AnF4


368

Molten Salt Reactor Fuel and Coolant

1400
1300
1200

T ( K)

1100
1000
900
800

700
600

0

0.2

0.4

0.6

0.8

1

x (ThF4)
Figure 7 The calculated BeF2–ThF4 phase diagram. Reproduced from van der Meer, J.; Konings, R. J. M.; Jacobs, M. H. G.;
Oonk, H. A. J. J. Nucl. Mater. 2005, 344, 94–99.
1600
1000

1400

T ( K)

900
800
700
600
500

0.0

0.0

0.1

T ( K)

1200

0.1
x (UF4)

0.2

0.2

1000

800

600
0.0

0.2

0.4

0.6


0.8

1.0

x (ThF4)

Figure 8 The equilibrium diagram of the LiF–ThF4 system assessed in Benesˇ et al.49: ○ thermal analysis data obtained
by Thoma et al.35; □ supercooled data;  invariant equilibria as reported in Thoma et al.35 Inset: calculated ThF4–UF4
pseudobinary system with constant amount of LiF at 78 mol%. Reproduced from Benesˇ, O.; Beilmann, M.; Konings, R. J. M.
J. Nucl. Mater. 2010, 405, 186–198.

basis of thermal analysis and thermal quenching.
Based on their data, the phase diagram was thermodynamically assessed by van der Meer et al.24 and more
recently by Benesˇ et al.49 The phase diagram from the

latter study,24 is shown in Figure 8. The LiF–ThF4 phase
diagram consists of four mixed compounds: Li3ThF7,
which melts congruently and Li7Th6F31, LiTh2F9, and
LiTh4F17, all melting peritectically. Two eutectic points


Molten Salt Reactor Fuel and Coolant

were found at xeut1 ¼ ð22:4 Æ 1Þmol% ThF4 with
Teut1 ¼ ð841 Æ 1ÞK, and xeut2 ¼ ð28:3 Æ 1Þmol%
ThF4 with Teut2 ¼ ð838 Æ 1ÞK, the first selected as a
fuel composition of the MSFR concept.
In this notation, AnF4 is represented mainly
by ThF4, which serves as a fertile material, and by
UF4, which is the fissile material, normally presented

with a concentration of up to 4 mol%. As UF4 and
ThF4 form close-to-ideal solid and liquid solutions,
the melting point of the fuel is only slightly affected
by the UF4/ThF4 substitution. The effect of UF4 addition is demonstrated in the inset graph of Figure 8,
which shows the calculated liquidus line (the very
upper line) of the ThF4–UF4 pseudobinary system
with the amount of LiF constant at 78 mol%. The
left axis of the graph corresponds to the proposed
LiF–ThF4 (78–22 mol%) fuel composition (eutectic1
of the LiF–ThF4 system) and the right axis corresponds to the LiF–UF4 (78–22 mol%) composition;
thus, in this case, all ThF4 is substituted by UF4.
As can be seen from the figure, the liquidus line along
this section is nearly constant, with a total drop of
only 18 K.
3.13.4.2.7 LiF–BeF2–AnF4

The LiF–BeF2–ThF4 system is the reference salt for
a MSR designed as a thermal breeder. The equilibrium diagram of this system was measured by

369

Thoma et al.34 It contains a single eutectic at
1.5 mol% ThF4 and Teut ¼ ð629 Æ 3ÞK; no ternary
compounds were found. van der Meer et al.36 calculated the ternary from the assessed binaries and found
good agreement with the experimental diagram. The
calculated phase diagram of the LiF–BeF2–ThF4
system is shown in Figure 9, as a projection of the
liquidus surface.
In the MSBR concept, the proposed fuel
composition in the LiF–BeF2–AnF4 system was 71.7–

16.0–12.3, where the AnF4 fraction was made up of
12.0 mol% ThF4 and 0.3 mol% UF4. In this section,
AnF4 is represented by pure ThF4, which is possible for
the same reasons as discussed in Section 3.13.3.1.1.
If we then assume that the concentration of ThF4
must be 12.3 mol%, it is possible, according to thermodynamic data, to determine the lowest melting
temperature of such a system and its exact composition.
It has been found at T ¼ 786 K and LiF–BeF2–ThF4
(67.1–20.6–12.3 mol%) (Composition 1), thus reasonably close to the data of the MSBR fuel (T ¼ 771 K and
LiF–BeF2–AnF4 (71.7–16.0–12.3 mol%) (Composition
2)). This means that, keeping the safety margin of 50 K,
the inlet temperature of the reactor must be a minimum of 836 K. This is a promising result because it
is lower than the inlet temperature in MSBR, which
was found to be 839 K, as discussed in Section 3.13.1.
According to the modeled phase diagram (Figure 9),

ThF4

13

00

12

00

11

00


10

00

90

0

90
0
00

10
11

00

LiF

BeF2

Figure 9 Calculated liquid surface of the LiF–BeF2–ThF4 phase diagram. Isotherms are labeled in K with interval of 25 K.
Reproduced from van der Meer, J.; Konings, R. J. M.; Oonk, H. A. J. J. Nucl. Mater. 2006, 357, 48–57.


370

Molten Salt Reactor Fuel and Coolant

the MOSART reactor.41 Note here that, in order

to simplify the study, all actinides were represented
by plutonium. This was possible as plutonium is the
major constituent of all actinides considered in
the MOSART fuel. A pseudoternary phase diagram
of the LiF–NaF–BeF2–(PuF3 ¼ 1.3 mol%) system is
shown in Figure 10. The melting temperature of
the lowest eutectic composition is calculated at
775 K, which is much lower than the designed inlet
temperature of the MOSART concept6 and therefore
acceptable for reactor purposes.
The optimized fuel composition as found in
Benesˇ and Konings25 varies slightly from that of
the MOSART concept (LiF–NaF–BeF2–PuF3 (14.8–
57.4–26.5–1.3)). Because the authors of the MOSART
concept did not have a full thermodynamic description
of the whole LiF–NaF–BeF2–PuF3 system, they took
the eutectic of the LiF–NaF–BeF2 system with the
lowest BeF2 content, as reported in Thoma,42 and
directly dissolved 1.3 mol% of AnF3 in it. Hence, they
did not consider the shift of the eutectic composition
while adding AnF3, which was demonstrated in Benesˇ
and Konings.25

the calculated liquidus temperature of the MSBR composition (Composition 2) is 795 K.
As the melting temperatures of Compositions 1 and
2 are very close, we focus (see Table 4) the discussion
only on the preferred composition of the MSBR concept (LiF–BeF2–ThF4 (71.7–16.0–12.3 mol%) (Composition 2)). This salt has also been more extensively
studied, and thus more of its properties are known.
3.13.4.2.8 LiF–NaF–BeF2–AnF3


The LiF–NaF–BeF2–PuF3 system is a reference salt in
the MOSART concept. The full thermodynamic
description of this quaternary system has been assessed
in a recent study by Benesˇ and Konings,25 using the
solubility data of PuF3 measured by Barton,37 Mailen
et al.,38 and Ignatiev et al.39,40 for the optimization of the
PuF3-containing ternary subsystems. Based on this
work,25 the optimized fuel composition is LiF–NaF–
BeF2–PuF3 (20.3–57.2–21.2–1.3), which is exactly
the point that corresponds to the lowest eutectic
in the LiF–NaF–BeF2–PuF3 system, with a fixed concentration of PuF3 at 1.3 mol% as an equilibrium
concentration of AnF3 after 10 years of operation of

BeF2
(805)
80
0

Miscibility gap

90
0

10

00

1015

1018


3)

(94

10

00

(89

2)

8)
(85
6)
(83

847

10

858

00

785
775

11


00

(1114)
LiF

(917)

(1263)
NaF

Figure 10 Calculated pseudoternary phase diagrams of the LiF–NaF–BeF2 system with constant amount of PuF3 ¼ 1.3 mol%.
Reproduced from Benesˇ, O.; Konings, R. J. M. J. Fluor. Chem. 2009, 130, 22–29.


Molten Salt Reactor Fuel and Coolant

371

1400

1200

T ( K)

1000
NaF + L
800

600


NaF + b-NaBF4
NaF + a-NaBF4

400

0

50

100

NaBF4 (mol%)
Figure 11 The equilibrium diagram of the NaF–NaBF4 system. Reproduced from Benesˇ, O.; Konings, R. J. M. J. Fluor.
Chem. 2009, 130, 22–29.

3.13.4.2.9 NaF–NaBF4

The equilibrium diagram of the NaF–NaBF4
system was studied by Selivanov and Stender,43 and
Barton et al.44 Both studies indicate that it is a simple
eutectic system, but the eutectic temperatures and
compositions differ considerably. In view of their
more careful sample preparation, the results of Barton
et al. are preferred, and this diagram is shown in
Figure 11. They found xeut ¼ (92 Æ 1) mol% NaBF4
with Teut ¼ (657 Æ 1) K.

T ¼ 727 K and LiF–NaF–KF (46.5–11.5–42.0 mol%).
Thermodynamic assessment of this system was done

in several studies,46–48 all of which were in close agreement. Figure 12 shows the LiF–NaF–KF phase diagram calculated using the data from the study by Benesˇ
and Konings,48 who found the ternary eutectic at
T ¼ 726 K and LiF–NaF–KF (45.3–13.2–41.5 mol%).
3.13.4.3 Solubility of Actinides in the
Fluoride Melt

3.13.4.2.10 LiF–NaF–KF

3.13.4.3.1 ThF4 in molten LiF

A eutectic mixture of LiF, NaF, and KF is one of the
possible candidates as an intermediate heat transfer
salt used to deliver the heat from the hightemperature reactor (advanced high-temperature
reactor (AHTR) or very high-temperature reactor
(VHTR)) to, for example, a hydrogen production
plant. Alternatively, the LiF–NaF–KF mixture can
be considered as a solvent for actinide trifluorides in
the molten salt actinide burner concept.
The LiF–NaF–KF phase diagram was measured by Bergmann and Dergunov,45 who found the
ternary eutectic with the lowest melting point at

The solubility of ThF4 in a matrix of LiF can be
deduced from the binary phase diagram in Figure 8.
For example, the solubility of ThF4 in a melt of LiF
for T ¼ 903 K (inlet temperature of the MSFR) is
between 20.0 and 32.3 mol%. Compositions in this
range are, thus, of interest as fuel for the MSFR.
In practice, the LiF–ThF4 (78–22 mol%) composition is the prime choice.
3.13.4.3.2 ThF4 in molten LiF–BeF2


The solubility of ThF4 in the LiF–BeF2 matrix has
been calculated for T ¼ 839 K (inlet temperature


372

Molten Salt Reactor Fuel and Coolant

NaF
(1269)

12

00

11

00

A

(991) 1000
10
00

1)

(92

B


11

00

C

726

(1131)
KF

(1120)
LiF

(763)

Figure 12 Calculated liquid surface of the LiF–NaF–KF phase diagram. Isotherms are labeled in K with interval of 25 K.
Primary phase fields: (A) (Li,Na,K)F; (B) (Na,K)F; (C) (Li,Na)F. Reproduced from Benesˇ, O.; Konings, R. J. M. J. Fluor. Chem.
2009, 130, 22–29.

LiF

0.9

0.1
0.3

0.7


A

0.2

0.8

C

B

0.6

0.5

0.5

0.4

Liquid

0.9

0.1

0.8

0.2

0.7


0.3

0.4

0.6

D
ThF4

0.9

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

BeF2

Figure 13 Isothermal plot of the LiF–BeF2–ThF4 phase diagram at T¼ 839 K. Reproduced from Benesˇ, O.;

Konings, R. J. M. J. Fluor. Chem. 2009, 130, 22–29.

of MSBR), keeping a constant ratio of LiF/BeF2 ¼
0.818/0.182. This ratio corresponds to the fuel composition proposed in MSBR. Figure 13 shows the
ternary phase diagram of the LiF–BeF2–ThF4 system

at T ¼ 839 K. The straight bold line represents the
LiF/BeF2 ratio at 0.818/0.182 within the whole field
of the diagram, while the ThF4 concentration varies
from 0 to 100 mol% as it moves from point ‘C’


Molten Salt Reactor Fuel and Coolant

towards ‘D.’ The solubility of ThF4 in the LiF–BeF2
matrix thus derived is between 9.2 and 20.8 mol%.
The interval of the solubility is represented by the ‘A’
and ‘B’ signs, respectively, which correspond to the
intersection of the ‘CD’ line with the surface of the
liquid field.
To our best knowledge, there are no experimental data
of the UF4 solubility in the LiF–ThF4 binary matrix.
However, based on the thermodynamic assessment
of the LiF–NaF–ThF4–UF4 system,49 the solubility
of UF4 in the LiF–ThF4 (78–22) composition (primary
fuel choice of the MSFR concept) has been calculated
for a temperature range of 840–880 K giving:
log10 Q ðmol%Þ ¼ 42:7475 À 0:1052T ðKÞ

log10 Q ðmol%Þ ¼ À4:0975 þ 4:32 Â 10À3 T ðKÞ


½2Š

According to the thermodynamic model of the LiF–
NaF–BeF2–PuF3 system published in Benesˇ and
Konings,25 the solubility of PuF3 in the recommended fuel matrix composition (LiF–NaF–BeF2
(20.6–57.9–21.5)) was calculated for the temperature
range of 823–973 K and fitted with the polynomial
equation below:
log10 Q ðmol%Þ ¼ À 5:3526 þ 9:7386 Â 10À3 T ðKÞ
À 3:4105 Â 10À6 T 2 ðKÞ

½1Š

3.13.4.3.4 PuF3 in molten LiF–BeF2

The solubility of PuF3 in the LiF–BeF2 binary melt
has been measured by Barton22 and Mailen et al.38
Barton measured the PuF3 solubility in LiF–BeF2
(71.3–28.7) and LiF–BeF2 (63–37) compositions
for the temperature range of 736–927 K, whereas
Mailen et al. measured the PuF3 solubility in LiF–
BeF2 (67–33) composition for the temperature range
of 59–657 K. Furthermore, Barton measured the
PuF3 solubility at T ¼ 838 K in the LiF–BeF2 matrix
as a function of composition from x(LiF) ¼ 0.52 to
0.72. Benesˇ and Konings25 recently evaluated the
LiF–NaF–BeF2–PuF3 system thermodynamically
and found very good agreement with all experimentally determined solubility data by Barton and Mailen
et al. On the basis of their assessment, the PuF3

Table 5

solubility in the LiF–BeF2 (67–33) composition has
been calculated for the temperature range of
780–930 K, giving:

3.13.4.3.5 PuF3 in molten LiF–NaF–BeF2

3.13.4.3.3 UF4 in molten LiF–ThF4

þ 6:6086 Â 10À5 T 2 ðKÞ

373

½3Š

Based on this equation, the total PuF3 solubility
in the LiF–NaF–BeF2 (20.6–57.9–21.5) melt at the
inlet temperature of the MOSART reactor concept
(T ¼ 873 K) is 3.55 mol%. This value is slightly
higher than the measured value in the MOSART
matrix composition (LiF–NaF–BeF2 (15–58–27)),
which was determined to be 3.08 mol%.6 Higher
solubility was achieved in the former case because
of the lower content of BeF2, which is the main fuel
component responsible for low AnF3 solubility, as
discussed in Benesˇ and Konings.25
3.13.4.3.6 PuF3 in molten LiF–BeF2–ThF4

The solubility of PuF3 in various compositions of

LiF–ThF4 and LiF–BeF2–ThF4 melts were measured
by Sood et al.,50 between 783 and 1060 K. Results
of their measurements are reported in Table 5,

Solubility of PuF3 in the LiF–BeF2–ThF4 melts measured by Sood et al.50

Salt composition (mol%)
LiF

BeF2

ThF4

74.0
76.9
75.3
68.2
71.6
71.3
70.0
75.0
80.0
75.0
70.0
65.0

22.1
17.1
16.7
20.5

16.2
15.5
14.0
5.0
0.0
0.0
0.0
0.0

3.9
6.0
8.0
11.3
12.2
13.2
16.0
20.0
20.0
25.0
30.0
35.0

Temperature
range (K)

A

ÀB Â 10À3

851–1021

878–973
812–1029
821–1049
796–1031
783–1060
802–949
826–1038
926–1054
882–1038
873–1018
935–1026

3.55 Æ 0.14
3.49 Æ 0.23
3.80 Æ 0.05
2.98 Æ 0.05
2.95 Æ 0.07
2.62 Æ 0.07
2.56 Æ 0.11
2.57 Æ 0.14
2.62 Æ 0.19
2.58 Æ 0.05
2.84 Æ 0.07
3.01 Æ 0.08

2.97 Æ 0.14
2.82 Æ 0.21
3.13 Æ 0.04
2.52 Æ 0.06
2.46 Æ 0.07

2.15 Æ 0.06
2.06 Æ 0.10
1.84 Æ 0.13
1.78 Æ 0.19
1.76 Æ 0.05
1.99 Æ 0.07
2.20 Æ 0.08


374

Molten Salt Reactor Fuel and Coolant

showing the derived coefficients for the general
equation:
log10 Q ðmol%Þ ¼ A þ B=T ðKÞ
3.13.4.4

½4Š

The agreement between the studies is excellent, as
shown in Figure 14 in an isothermal section at
873 K. From the results, we interpolate for the 66–34
composition:
ðmPa sÞ ¼ 0:116expð3755=T ðKÞÞ

Density and Viscosity

3.13.4.4.1 LiF–BeF2


The density of liquid LiF–BeF2 has been measured
by Blanke et al.51 from 0 to 55 mol% BeF2, by Cantor
et al.52 for 50.2, 74.9, and 89.2 mol% BeF2, and by
Cantor53 for the 34 mol% BeF2 composition. As discussed by van der Meer et al.,36 the molar volumes
derived from the measured density data indicate
ideal behavior, suggesting that the density can be
interpolated from the molar volume data for the
pure components. However, the density and molar
volume of liquid BeF2 are known only at a single
temperature (T ¼ 1073 K), and not at all as a function
of temperature. Therefore, we have selected the
results of the 66:34 composition from Cantor53:
rðkg mÀ3 Þ ¼ 2146:3 À 0:4884T ðKÞ

½5Š

The viscosity of liquid LiF–BeF2 has been measured
by Cohen and Jones54 and Abe et al.55 for the compositions 31and 32.8 mol% BeF2, respectively, as well as
by Blanke et al.,51 Cantor et al.,52 and Desyatnik et al.56
for a wide(r) range of compositions and temperatures.

5

3.13.4.4.2 LiF–AnF4

The density of LiF–ThF4 mixtures was measured by
Porter and Meaker57 and Hill et al.58 The data are in
good agreement and clearly indicate a linear dependence of the molar volume on composition as shown
in Figure 15, confirming ideal behavior. The density
of the 78–22 composition as measured by Porter and

Meaker,57 is given by:
rðkg mÀ3 Þ ¼ 5543 À 1:25T ðKÞ

½7Š

The density of liquid LiF–UF4 mixtures were
measured by Blanke et al.51 and Porter and Meaker.57
The results are in excellent agreement, as shown in
Figure 15. The results indicate a linear dependence
of the molar volume on composition, confirming
ideal behavior.
The viscosity of LiF–ThF4 mixtures was measured
by Chervinskij et al.59 from 0 to 100 mol% ThF4. The
results reveal a strong deviation from ideal behavior
around the eutectic composition. An isothermal section in Figure 15 shows a steady increase from LiF
to ThF4. The viscosity of the 78–22 composition
interpolated from the results is given by:
ðmPa sÞ ¼ 0:365expð2735=T ðKÞÞ

½8Š

The viscosity of the LiF–UF4 system measured by the
same group60 shows a less strong increase with the AnF4
content compared to ThF4 (Figure 15). As a result, the
above equation probably overestimates the viscosity
slightly in the case of part replacement of ThF4 by UF4.

3
log h (Pa s)


½6Š

1

A

3.13.4.4.3 LiF–BeF2–ThF4
-1
B
-3
0.00

0.20

0.40
0.60
x (BeF2)

0.80

1.00

Figure 14 The viscosity of liquid LiF–BeF2 at 873 K:
▽ Cohen and Jones54; △ Blanke et al.51; □ (curve A),
Cantor et al.52; ○ (curve B), Desyatnik et al.56;
◊ Abe et al.55 Reproduced from Benesˇ, O.; Konings, R. J. M.
J. Fluor. Chem. 2009, 130, 22–29.

The densities of the three compositions of the
LiF–BeF2–ThF4 system, with almost constant

LiF concentration, were measured by Cantor.53
Unfortunately, the density of the LiF–BeF2–ThF4
(71.7–16.0–12.3) composition has not been measured;
however, a very close composition (LiF–BeF2–ThF4
(70.06–17.96–11.98)) has been determined and the
corresponding density function is given below:
rðkg mÀ3 Þ ¼ 4043:9 À 0:8064T ðKÞ

½9Š

It has been shown by van der Meer and Konings61 that
the molar volumes and thus the densities of all three


Molten Salt Reactor Fuel and Coolant

50

25
An = Th

20

An = U

h (mPa s)

Vm (cm3 mol−1)

40


30

20

10
0.00

375

15

An = Th

10

An = U

5

0.20

0.40 0.60
x (AnF4)

0.80

1.00

0

0.00

0.20

0.40 0.60
x (AnF4)

0.80 1.00

Figure 15 The molar volume (left) and viscosity (right) of liquid LiF–ThF4 and LiF–UF4 at 1273 K. Right figure: ○ data by
Chervinskij et al.59; □ data by Desyatnik et al.60; Left figure: ▲ data by Hill et al.58; ▼ data by Porter and Meaker57; □ data
by Blanke et al.51; ○ data by Porter and Meaker.57 Reproduced from Benesˇ, O.; Konings, R. J. M. J. Fluor. Chem. 2009,
130, 22–29.

LiF–BeF2–ThF4 compositions measured in Cantor53
behave almost ideally. Based on this triplet of data and
with the assumption of the ideality, it is possible to
estimate the density function of temperature of pure
BeF2, which has not been measured yet. The density of
liquid BeF2 was measured by Mackenzie,62 but only
at 1073 K, obtaining the value of 1947 Æ 10 kg mÀ3.
Cantor et al.52 also measured the density, but, due to
the experimental difficulties, they derived only an
approximate value: 1960 kg mÀ3 at 1123 K. The value
of MacKenzie is recommended and taken as a constraint
in our estimation. The obtained density for liquid BeF2
as a function of temperature is shown below:
rðkg mÀ3 Þ ¼ 3190:5 À 1:1589T ðKÞ

½10Š


Using eqn [10] together with the selected data for
the LiF and ThF4 densities taken from van der
Meer and Konings,61 we have calculated the expected
density function of temperature for the LiF–BeF2–
ThF4 (71.7–16.0–12.3 mol%) composition (MSBR).
The obtained equation is given below:
rðkg mÀ3 Þ ¼ 4124:3 À 0:8690T ðKÞ

½11Š

The results from eqns [9] and [11] agree very well.
As the former equation is based on the experimental
results whereas the latter is an estimate, and both
equations refer to very similar compositions, the
extrapolation of the density in the LiF–BeF2–ThF4
system can be justified on the basis of ideal behavior.
Based on eqn [11], the density of the salt mixture at
T ¼ 973 K is 3279 kg mÀ3, for the LiF–BeF2–ThF4–
UF4 (71.0–16.0–12.0–1.0 mol%) composition, while the

reported density at the same temperature taken from the
study by Briant and Weinberg63 is 3250 kg mÀ3: values
that are in close agreement.
The viscosity of liquid LiF–BeF2–ThF4 of two
compositions was measured by Cantor.53 The viscosity of the quaternary LiF–BeF2–ThF4–UF4 (71–16–
12–1) composition, which is nearly identical to
our reference selection (LiF–BeF2–ThF4 (71.7–16–
12.3)), has been reported in Powers et al.64 for the
temperature range of 873–1073 K, giving:

ðmPasÞ ¼ 0:062expð4636=T ðKÞÞ

½12Š

3.13.4.4.4 LiF–NaF–BeF2–AnF4

Densities of several LiF–NaF–BeF2 mixtures have
been measured in various studies,6,64 but the exact
compositions are different from that of our recommended fuel choice from Table 3. However, in a
recent study by Khokhlov et al.65 the density of
a very similar ternary mixture (LiF–NaF–BeF2
(22–56.7–21.3 mol%)) was estimated, using an additive P
law of molar volumes according to the equation
V¼
NiVi, where Vi is a molar volume of LiF and
NaF end members, and LiF–BeF2 and NaF–BeF2 mixtures, whose compositions are shown in square brackets
in the following notations: [0.508LiF–0492BeF2]–
0.567NaF; [0.727NaF–0.273BeF2]–0.22LiF. The density
of the ternary mixture was taken as a mean value from
these two notations, and the temperature function thus
derived is shown below:
rðg cmÀ3 Þ ¼ 2:5777 À 5:38 Â 10À4 T ðKÞ

½13Š


Molten Salt Reactor Fuel and Coolant

Densities of binary LiF–BeF2 and NaF–BeF2 mixtures
were measured as a function of temperature and composition and taken from the work of Janz66 as reported

in Khokhlov et al.65 Khokhlov et al. also made calculations for the same compositions as measured by
Zherebtsov and Ignatiev6 (LiF–NaF–BeF2 (15–58–
27 mol%) and LiF–NaF–BeF2 (17–58–25 mol%)) and
in both cases found good agreement with the experimental data, which gave legitimacy to their approach.
Assuming that the density of the recommended
fuel matrix (LiF–NaF–BeF2 (20.6–57.9–21.5 mol%))
follows eqn [13], we can estimate the density of
the fuel with the contribution of 1.3 mol% PuF3,
using the additive law of molar volumes. For this
calculation, we need to know the molar volume of
pure liquid PuF3, which, to our best knowledge, has
not been determined experimentally. To derive this
quantity, we assume that liquid PuF3 has the same
molar volume as CeF3, which was obtained from the
density measured by Kirshenbaum and Cahill67 for
the temperature range of 1700–2200 K. For its similar
chemical behavior, CeF3 is considered as a proxy
compound to the plutonium species, a consideration
that is supported by the comparison of the ionic radii
of Ce3þ and Pu3þ, which are nearly identical, 115 pm
for Ce3þ and 114 pm for Pu3þ. The density function
of pure liquid PuF3 thus obtained is:
rðkg mÀ3 Þ ¼ 9550:6 À 1:4296T ðKÞ

½14Š

rðkg m Þ ¼ 2759:9 À 0:5730T ðKÞ

½15Š


for the fuel composition (LiF–NaF–BeF2–PuF3
(20.3–57.2–21.2–1.3)).
To estimate viscosity, Khokhlov et al.65 applied a
similar approach as in the case of density. According
to their report, the input data were the experimental
results of the molar viscosities of binary LiF–BeF2,
NaF–BeF2,56 and LiF–NaF melts.68 The obtained
temperature function of kinematic viscosity of the
LiF–NaF–BeF2 (22–56.7–21.3 mol%) composition
is shown in Figure 16. The same figure shows a
comparison of the estimated curve with the experimental data measured by Ignatiev et al.69,70 and
there is a close agreement between both sets of
results. The corresponding dynamic viscosity of the
LiF–NaF–BeF2 (22–56.7–21.3 mol%) composition is
given in the following equation:
log10 ðmPa sÞ ¼ À1:0018 þ ð1617:4=T ðKÞÞ

3

2

1

0

850

900

950


1000

½16Š

As this composition is very close to the recommended
fuel choice, neglecting the influence of addition of a

1050

T ( K)
Figure 16 Kinematic viscosity of the LiF–NaF–BeF2
(22–56.7–21.3 mol%) melt: (——) estimated data from
Khokhlov et al.65; (□) experimental data by Ignatiev et al.69,70
Reproduced from Benesˇ, O.; Konings, R. J. M. J. Fluor.
Chem. 2009, 130, 22–29.

relatively small amount of PuF3 (1.3 mol%), we recommend eqn [16] as a viscosity function of the LiF–
NaF–BeF2–PuF3 (20.3–57.1–21.2–1.3 mol%) fuel.
3.13.4.4.5 NaF–NaBF4

The density of NaF–NaBF4 (8–92 mol%) was
measured by Cantor53 from 673 to 864 K. The results
can be represented by the equation:
rðkg mÀ3 Þ ¼ 2446:3 À 0:711T ðKÞ

giving
À3

4


n ϫ 106 (m2 s−1)

376

½17Š

The viscosity of NaF–NaBF4 (8–92 mol%) was
measured by Cantor53 from 682 to 810 K. The results
can be represented by the equation:
ðmPa sÞ ¼ 0:0877expð2240=T ðKÞÞ

½18Š

3.13.4.4.6 LiF–NaF–KF

The density of the eutectic melt of the LiF–NaF–KF
system has been measured by Chrenkova´ et al.71
for the temperature range of 940–1170 K. The exact
composition of the LiF–NaF–KF melt measured
in their study was x(LiF) ¼ 0.465, x(NaF) ¼ 0.115,
and x(KF) ¼ 0.420, thus corresponding to the eutectic
composition found by Bergmann and Dergunov.45
The density as a function of temperature of the eutectic
composition has also been reported by Powers et al.64
for an unspecified temperature range. As shown in
Figure 17, the data by Chrenkova´ et al. and Powers
et al. differ significantly. The results of Chrenkova´ et al.
are close to the density that is calculated assuming
ideal behavior and the curve has almost the same



Molten Salt Reactor Fuel and Coolant

377

2100

1.0

2050

0.9
Density

2000

0.8
0.7

1900

0.6

1850
1800

0.5

r (kg m–3)


log10 h (mPa s)

1950

Viscosity
1750

0.4
1700
0.3
0.2
750

1650
800

850

900

950

1000

1050

1100

1150


1600
1200

T ( K)
Figure 17 Viscosity and density functions of temperature reported by Chrenkova´ et al.71 (– – –) and Powers et al.64 (——).
For comparison, the ideal density behavior is represented by a dotted line. Reproduced from Benesˇ, O.; Konings, R. J. M.
J. Fluor. Chem. 2009, 130, 22–29.

slope, which is consistent with our observations that
most of these fluoride systems are ideal. For this reason,
we recommend the data by Chrenkova´ et al.:
rðkg mÀ3 Þ ¼ 2579:3 À 0:6240T ðKÞ

½19Š

The viscosity of the eutectic melt of the LiF–NaF–KF
system has been measured by Chrenkova´ et al.71 for
the temperature range of 773–973 K and Powers et al.64
for the temperature range of 873–1073 K. The comparison between the data by Chrenkova´ et al. and
by Powers et al. is shown in Figure 17. The data by
Chrenkova´ et al. have been selected:
log10 ðmPa sÞ ¼ À1:6044 þ 1944=T ðKÞ

½20Š

3.13.4.5 Heat Capacity and Thermal
Conductivity
3.13.4.5.1 LiF–BeF2


The heat capacity of liquid LiF–BeF2 (66–34 mol%)
has been measured by Hoffman and Cooke (as cited
in Cantor et al.72), and Douglas and Payne,73 who
obtained 2.41 J KÀ1 gÀ1 (unknown temperature range)
and 2.37 J KÀ1 gÀ1 (773–873 K), respectively. The
value Cp(LiF–BeF2 (66–34 mol%)) ¼ 2.39 J KÀ1 gÀ1
has been selected.
The thermal conductivityof LiF–BeF2 (66–34 mol%)
has been measured by Cooke (as reported in Cantor
et al.72) to be 1.0 W mÀ1 KÀ1, independent of the

temperature. Some time later, Cooke et al.74 reported
more detailed results, indicating that the thermal conductivity increases slightly, from l ¼ 1.0 W mÀ1 KÀ1
at 923 K, to about 1.2 W mÀ1 KÀ1 between 1023 and
1133 K. Kato et al.75 measured the thermal diffusivity
of the compositions 66–34 mol% and 53–47 mol%.
From their results, we calculate 1.1 W mÀ1 KÀ1 for
the 66–34 mol% composition, which is in good agreement with Cooke’s results, and we recommend l
(LiF–BeF2 (66–34)) ¼ 1.1 W mÀ1 KÀ1.
3.13.4.5.2 LiF–AnF4

To our best knowledge, heat capacity or thermal
conductivity have not been measured for the LiF–
ThF4 system. We have estimated the heat capacity of
the LiF–ThF4 (78–22 mol%) composition on the
basis of the comparison between the ideal heat capacity and the measured data from other fluoride systems taken from Powers et al.64 The average positive
deviation from the ideal behavior has been found to
be 11%. If we combine this difference with the ideal
heat capacity of the LiF–ThF4 (78–22 mol%) composition,
we

obtain
our
suggested
value:
Cp ¼ 1.0 J gÀ1 KÀ1.
There are not enough data to accurately estimate the thermal conductivity of the LiF–ThF4
(78–22 mol%) composition; however, we suggest
that the value be slightly higher than the value of the
LiF–BeF2 (66–34 mol%) composition and close to


378

Molten Salt Reactor Fuel and Coolant

the value for LiF–BeF2–ThF4 (71.7–16–12.3 mol%)
composition, which was derived for T ¼ 1023 K
(see Section 3.13.4.5.3). Our suggested value
for LiF–ThF4 (78–22 mol%) composition is
l ¼ $1.5 W mÀ1 KÀ1.
3.13.4.5.3 LiF–BeF2–ThF4

Araki and Kato76 measured the thermal diffusivity of
liquid LiF–BeF2–ThF4 (64–18–18 mol%), from
which they derived the thermal conductivity using
their heat capacity data and an estimated density. The
results indicate an almost constant value in the temperature range of 850–1000 K: 0.95–0.98 W mÀ1 KÀ1.
The recommended heat capacity according to
Araki and Kato is Cp ¼ 1.23 J gÀ1 KÀ1. Both data,
heat capacity and thermal conductivity, are measured

for a LiF–BeF2–ThF4 composition that is slightly
different from the one considered in this work
(71.7–16.0–12.3 mol%). Cooke et al.74 reported (in
graphical form only) the thermal conductivity of
liquid LiF–BeF2–ThF4–UF4 (67.5–20–12–0.5 mol%)
for the temperature range of 800–1150 K. The data
scatter around l ¼ 1.2–1.4 W mÀ1 KÀ1, with a suggested maximum at 973 K. This result is somewhat
different from that of Araki and Kato.76 As the results
for liquid LiF–BeF2 from both groups are in good
agreement, the variation probably arises from differences in BeF2 and MF4 content (where M ¼ Th, U,
and Zr). The results from the above-mentioned

sources74,76 indicate that in the measured composition
range, the thermal conductivity decreases with increasing (BeF2 þ MF4) content as indicated in Figure 18.
The LiF–BeF2–ThF4 (71.7–16.0–12.3 mol%) composition is just outside this range (x(BeF2 þ MF4) ¼
28.3 mol%), and linear extrapolation would suggest
l ¼ 1.51 W mÀ1 KÀ1 at T ¼ 1023 K (solid line in
Figure 18). However, such linear extrapolation
would suggest a relatively high thermal conductivity
of LiF–ThF4 (78–22 mol%). Alternatively, one could
extrapolate the results in a nonlinear way (dashed line
in Figure 18). This would suggest l ¼ 1.49 W mÀ1
KÀ1 at T ¼ 1023 K, which is very close to previously
established value. In this case, the thermal conductivity
of LiF–ThF4 (78–22 mol%) is 1.6 W mÀ1 KÀ1, which is
more realistic. For LiF–BeF2–ThF4 (71.7–16.0–12.3)
composition we recommend:
l ¼ 1:5 WmÀ1 KÀ1

The heat capacity of the quaternary LiF–BeF2–ThF4–

UF4 (71–16–12–1 mol%) composition, which is
nearly identical to our reference selection (LiF–BeF2–
ThF4 (71.7–16–12.3 mol%)), has been reported in
Briant and Weinberg,63 giving Cp ¼ 1550 J kgÀ1 KÀ1.
This value is also fairly close to the estimated value,
based on the approach published by Khokhlov et al.65
(discussed in the following section), which gives
Cp ¼ 1.506 J gÀ1 KÀ1. We select the measured value,
Cp ¼ 1.550 J gÀ1 KÀ1.

1.6

l ( W m-1 K-1)

1.4

1.2

1.0

0.8
0.28

T = 1023 K

0.30

½21Š

0.32

x (BeF2 + MF4)

0.34

0.36

Figure 18 Extrapolation of the thermal conductivity of the LiF–BeF2–ThF4 (71.7–16.0–12.3 mol%) composition at
T ¼ 1023 K. (——) linear fit; (– – –) polynomial fit. (▪) Experimental data from Cooke et al.74 and Araki and Kato.76
Reproduced from Benesˇ, O.; Konings, R. J. M. J. Fluor. Chem. 2009, 130, 22–29.


Molten Salt Reactor Fuel and Coolant

3.13.4.5.4 LiF–NaF–BeF2–PuF3

Because of the lack of experimental data on the heat
capacity of the actinide-containing salts, it is difficult
to properly assess the value for the LiF–NaF–BeF2–
PuF3 (20.3–57.1–21.2–1.3 mol%) composition. However, Khokhlov et al.65 recently evaluated the heat
capacity of more than 30 fluoride salts and found a
simple empirical dependence on the inverse molar
mass (1/M) by the following equation:
Cp ð J KÀ1 gÀ1 Þ ¼ 0:2916 þ 0:00802Â104 =M

½22Š

Using the above equation, the heat capacity for
the fuel composition from Table 3 is calculated as
2.15 J KÀ1 gÀ1. This value is fairly close to the experimentally determined heat capacity of the plutoniumfree LiF–NaF–BeF2 (24–53–23 mol%) composition,
which was found at 2.26 J KÀ1 gÀ1. Because this composition is similar to the fuel composition and its heat

capacity is only slightly higher than that found for
the fuel composition using eqn [22], we recommend
2.15 J KÀ1 gÀ1 as a reasonable estimate of the heat
capacity.
Because of the lack of experimental data, it is difficult
to assess the thermal conductivity of the complicated
salt mixtures, such as plutonium-containing fuel; however, Khokhlov et al.65 analyzed the experimental values
of the thermal conductivity determined earlier for molten chlorides, bromides, and iodides of alkali metals and
their mixtures and deduced an equation describing the
experimental results within the measurement errors.
The obtained equation depends only on temperature
T (expressed in K) and the molar weight M of the salt
mixture (expressed in g molÀ1) and is given by:
À1

À1

379

function of the thermal conductivity has been determined by a linear fit, giving:
lðWmÀ1 KÀ1 Þ ¼ 0:66À2:37Â10À4 T ðKÞ

½25Š

It is interesting to compare these results with those
of Cantor et al.,72 who reported preliminary measurements of the thermal conductivity of pure liquid
NaBF4, finding l ¼ 0.51 W mÀ1 KÀ1, which is, on average, slightly higher than that of the NaF–NaBF4
(8–92 mol%) eutectic composition.
3.13.4.5.6 LiF–NaF–KF


Powers et al.64 reported the heat capacity of the
LiF–NaF–KF (46.5–11.5–42 mol%) melt measured
at T ¼ 973 K, giving Cp ¼ 1.88 J gÀ1 KÀ1. This value
is significantly higher than that obtained from the
ideal behavior (Cp, ideal ¼ 1.66 J gÀ1 KÀ1).
The same authors measured the thermal conductivity of the eutectic composition, giving l ¼ 4.5
W mÀ1 KÀ1. This value is much higher than the
measurement (773–1173 K) by Ewing et al., l ¼ 0.6
W mÀ1 KÀ1. Smirnov et al.77 measured the thermal
conductivity of eutectic LiF–NaF–KF (46.5–11.5–
42 mol%) from 790 to 1080 K and obtained l ¼ 0.36 þ
5.6 Â 10À4T(K) W mÀ1 KÀ1, giving 0.8 W mÀ1 KÀ1
at T ¼ 773 K. Kato et al.75 measured the thermal
diffusivity of LiF–NaF–KF (46.5–11.5–42 mol%) in
the temperature range of 730–823 K and obtained
a ¼ 7.6 Â 10À4 þ 6.3 Â 10À7T (K) m2 hÀ1, which yields
0.8 W mÀ1 KÀ1 at T ¼ 773 K when combined with the
selected heat capacity and density values. We thus
recommend:
lðWmÀ1 KÀ1 Þ ¼ 0:36 þ 5:6Â10À4 T ðKÞ

½26Š

À3

lðWm K Þ ¼ À0:34 þ 0:5  10 T þ 32:0 =M ½23Š

Using this equation, the thermal conductivity of the
LiF–NaF–BeF2–PuF3 (20.3–57.1–21.2–1.3) composition gives the following function of temperature:
À1


À1

À3

lðWm K Þ ¼ 0:402 þ 0:5 Â 10 T

½24Š

3.13.4.5.5 NaF–NaBF4

The heat capacity of the NaF–NaBF4 (8–92 mol%)
melt has been determined by Dworkin (as mentioned
in Cantor53) as Cp ¼ 1.506 J gÀ1 KÀ1.
The thermal conductivity of the NaF–NaBF4
(8–92 mol%) melt has been reported by Cooke
et al.74 for the temperature range of 740–1000 K.
However, they have reported their results only in a
graphical form without listing the exact values or equations. Thus, their data have been obtained by digital
subtraction from the figure, and the temperature

3.13.4.6

Vapor Pressure

3.13.4.6.1 LiF–BeF2

According to the thermodynamic data taken from
Benesˇ and Konings,25 the vapor pressure of the
LiF–BeF2 (66–34 mol%) composition has been calculated for the temperature range between 823 and

1473 K, which covers the typical operating temperature range of the MSR and also describes the vapor
pressure at high temperature in order to simulate the
fuel behavior during accidental conditions. The
result is given in the equation below:
log10 pðPaÞ ¼ 11:914À13 003=T ðKÞ

½27Š

3.13.4.6.2 LiF–AnF4

According to the thermodynamic data obtained
from van der Meer et al.,36 the vapor pressure of the


380

Molten Salt Reactor Fuel and Coolant

LiF–ThF4 (78–22 mol%) composition has been calculated for the temperature range between 839 and
1473 K. The result is given in the equation below:
log10 pðPaÞ ¼ 11:902À12 989=T ðKÞ

where the total vapor pressure is highlighted by a
bold curve, whereas the most volatile species are
reported by thin lines. The graph does not include
Pu containing species because even the most volatile
among these, PuF4, has a much lower pressure than
the species reported, and therefore they have been
excluded from the figure. The total vapor pressure is
represented by the following equation:


½28Š

The vapor pressure of the LiF–ThF4–UF4 (78–18–
4 mol%) composition is slightly lower compared to a
system with no UF4 content. The calculated boiling
temperature of the LiF–ThF4 (78–22 mol%) composition is T ¼ 1874 K.

log10 pðPaÞ ¼ 11:6509 À 12 827=T ðKÞ

which gives p ¼ 0.001 Pa and p ¼ 0.046 Pa at the
designed inlet temperature (Tinlet ¼ 873 K) and the
outlet temperature (Toutlet ¼ 988 K) of the MOSART
reactor,6 respectively. Both values are very low, and
hence the composition shift of the fuel as a consequence
of the incongruent vaporization can be neglected.
The calculated boiling temperature is T ¼ 1973 K.

3.13.4.6.3 LiF–BeF2–ThF4

According to the thermodynamic data by van der
Meer et al.,36 the vapor pressure of the LiF–BeF2–
ThF4 (71.7–16.0–12.3 mol%) composition has been
calculated for the temperature range of 823–1473 K
and the obtained result is shown in the following
equation:
log10 pðPaÞ ¼ 11:158 À10 790:5=T ðKÞ

½30Š


3.13.4.6.5 NaF–NaBF4

½29Š

The vapor pressure of BF3 in the NaF–NaBF4 system
has been measured by Cantor et al.78 They measured
the equilibrium of the BF3 gaseous species over the
melt for the composition range of 5–100 mol%
NaBF4 and the temperature range of 698–1473 K.
However, in their report they ‘only’ show the results
for 900, 1000, and 1100 K. Based on this triplet of
data, the vapor pressure equation of NaF–NaBF4
(8–92 mol%) has been determined, giving:

The calculated boiling temperature of the LiF–
BeF2–ThF4 (71.7–16.0–12.3 mol%) composition is
T ¼ 1744 K.
3.13.4.6.4 LiF–NaF–BeF2–AnF3

In the study by Benesˇ and Konings,25 the vapor pressure of the potential fuel composition (LiF–NaF–
BeF2–PuF3 (20.3–57.1–21.2–1.3 mol%)) has been
calculated, and the results are reported in Figure 19,

log10 pðPaÞ ¼ 11:638À6550:6=T ðKÞ

½31Š

1E - 4
tal
To


pvapor (atm)

1E - 5
F2
Na 2

1E - 6

F
Be
1E - 7

eF

F

Li

1E - 8

1E - 9

2

F2
Li 2
F3
Li 3


3

LiB

F

Na

900

1000

1100
T ( K)

1200

1300

Figure 19 Calculated vapor pressure of the x(LiF) ¼ 0.203, x(NaF) ¼ 0.571, x(BeF2) ¼ 0.212, x(PuF3) ¼ 0.013 potential
fuel composition. Reproduced from Benesˇ, O.; Konings, R. J. M. J. Chem. Thermodyn. 2009, 41, 1086–1095.


Molten Salt Reactor Fuel and Coolant

3.13.4.6.6 LiF–NaF–KF

The vapor pressure of the LiF–NaF–KF (46.5–11.5–
42 mol%) composition has been calculated for
the temperature range between 823 and 1473 K in

a study by Benesˇ and Konings,79 on the basis of
the thermodynamic data taken from Benesˇ and
Konings.48 The result is given by the equation below:

381

this oxide is very insoluble in the fluoride mixture
of the MSBR composition given by:
log10 QPa2 O5 ¼ 0:91À12 760=T ðKÞ

½38Š

where
5=2

QPa2 O5 ¼ xPa5þ xO2À

½39Š

log10 QThO2 ¼ À2:86 À 3280=T ðKÞ

½33Š

log10 QPaO2 ¼ À2:86 À 4920=T ðKÞ

½34Š

Whether Pa2O5 will precipitate or not depends on
three factors: oxide and protactinium concentrations,
and the oxidation state of the fuel, which, in the MSR,

is controlled by the UF4/UF3 ratio, as discussed in
Section 3.13.8. As reported in Rosenthal et al.,80 with
100 ppm Pa and 30 ppm oxide present, the UF4/UF3
ratio must be at least 105 in order to start the Pa2O5
precipitation. Nevertheless, such oxidizing conditions
are easily avoided, as the typical UF4/UF3 ratio in the
MSR is set to 100 (see Section 3.13.8).
Even stronger oxidizing conditions (UF4/UF3 > 108)
are required to precipitate PuO2, and hence this species
is avoided in the MSR fuel as well.
Although the Pa2O5 and PuO2 species will
not be formed in the fuel salt, the other actinide
dioxides UO2, ThO2, and PaO2 can be formed
under the redox conditions of the MSR and, due
to the very low solubilities of these species in the
fluoride matrix (as given by eqns [33]–[36]), they can
easily precipitate in the solid form. Therefore, it is
important to keep the fuel salt free from any oxide
contamination to avoid this inadvertent event. This
will certainly require some care but, as mentioned in
Rosenthal et al.,80 the results of the MSRE project
have shown that the oxide content can be maintained
at an adequately low level in order to achieve successful long-term operation of the MSR.

log10 QUO2 ¼ À2:86 À 5660=T ðKÞ

½35Š

3.13.6 Electroanalytical Chemistry


log10 QPuO2 ¼ À2:86 À 7100=T ðKÞ

½36Š

log10 pðPaÞ ¼ 10:748À10 789=T ðKÞ

½32Š

3.13.5 Role of Oxygen Impurities
In the previous section, the physicochemical properties of pure fluoride salts have been discussed.
However, the behavior of these systems can be significantly affected by the presence of the oxide
ion that might be resulting from contamination of
the salt system; for example, the presence of reactive
oxides such as H2O can result in precipitation of
the UO2 phase.80 Therefore, the effect of added
oxide on the fuel mixture containing LiF, BeF2,
ThF4, UF4, and PaF4 has been investigated in several studies,81–88 as reported in Rosenthal et al.80 who
give a summary of the main conclusions from these
works is given. It has been found that the solubility
of the actinide dioxides in the MSBR fuel salt is low
and it decreases in the order, ThO2, PaO2, UO2, and
PuO2. The temperature functions of the solubilities
of these oxides were estimated in the same study as
follows:

where
QMO2 ¼ xM4þ xO2 2À

½37Š


The ThF4 concentration in the MSBR concept is
equal to x ¼ 0.12, and it has been shown80 that at
such concentrations of thorium, the ThO2 precipitation at T ¼ 773 K will start for xO2À ! 8Â10À4.
Protactinium is produced in thorium-containing
breeder fuel by neutron capture, and both tetravalent
and pentavalent species of protactinium are stable.
Thus, in addition to PaO2, Pa2O5 can precipitate
in the oxide form. As reported in Rosenthal et al.,80

Surprisingly, very little experimental work has been
done on the electrochemical properties of the main
ions in molten fluoride salts. For the LiF–BeF2 system,
some direct measurements of the standard potentials
have been made. The standard potentials of the main
ions in the liquid LiF–BeF2 (67–33) melt have been
reported by Baes.89–91 He has made an extensive analysis of the available literature, which is essentially
based on a comparative scale as only the Be2þ/Be0
couple has been measured electrochemically92:
BeðcrÞ þ 2HFðgÞ ¼ BeF2 ðslnÞ þ H2 ðgÞ

½IŠ

Using equilibrium constants, Gibbs energies of the
solutes, and activity coefficients, Baes derived the


382

Molten Salt Reactor Fuel and Coolant


Table 6
Standard potential in LiF–BeF2 (66–34) relative
to the HF(g)/H2 couple, E/V ¼ a þ bT (K)

Table 7
Standard potential in LiF–CaF2 (77–23) relative to
the F2/FÀ pair measured by Chamelot et al.93 at T ¼ 1100 K

Cell reaction

a

b  103

Cell reaction

E0/V

Liþ (sln) þ eÀ ¼ Li(cr)
Be2þ (sln) þ 2eÀ ¼ Be(cr)
1/2F2(g) þ eÀ ¼ FÀ(sln)
Th4þ(sln) þ 4eÀ ¼ Th(cr)
U3þ(sln) þ 3eÀ ¼ U(cr)
U4þ(sln) þ 4eÀ ¼ U(cr)
UF6(g) þ 2eÀ ¼ U4þ(sln) þ 6FÀ(sln)
Pu3þ(sln) þ 3eÀ ¼ Pu(cr)
Cr2þ(sln) þ 2eÀ ¼ Cr(cr)
Fe2þ(sln) þ 2eÀ ¼ Fe(cr)
Ni2þ(sln) þ 2eÀ ¼ Ni(cr)


À3.322
À2.460
þ2.827
À2.498
À2.059
À1.851
À1.439
À2.313
À0.898
À0.527
À0.357

0.763
0.694
0.044
0.720
0.626
0.807
0.200
0.788
0.508
0.516
0.830

Liþ(sln) þ 1eÀ ¼ Li(cr)
Th4þ(sln) þ 4eÀ ¼ Th(cr)
Nd3þ(sln) þ 3eÀ ¼ Nd(cr)
Gd3þ(sln) þ 3eÀ ¼ Gd(cr)

À5.33

À4.57
À4.88
À4.93

0.01
–0.30
–0.39

–1.05

E(V)

–1.36
–1.53
–1.77
–1.78

Fe2+/Fe
Ni2+/Ni
Cr2+/Cr
U4+/U3+
U4+/U
Pu3+/Pu
Be2+/Be
Th4+/Th

–2.56

Li+/Li


Table 8
Standard potential in LiF–CaF2 (77–23) relative
to the F2/FÀ pair measured by Hammel et al.94 at T ¼ 993 K
Cell reaction

E0/V

Liþ(sln) þ 1eÀ ¼ Li(cr)
U3þ(sln) þ 3eÀ ¼ U(cr)
U4þ(sln) þ 1eÀ ¼ U3þ(sln)

À5.44
À4.53
À3.81

salt, has a much narrower electrochemical window
and is not suitable for the reduction of the Th, Nd,
and Gd metals.
Hammel et al.94 measured the electrochemical
potential of UF4 in LiF–CaF2 (77–23) and found
UF4 less stable than the solvent components and
thus suitable for reduction from this salt. The values
of the redox potentials obtained in their study are
summarized in Table 8, showing the value for the
Liþ þ eÀ ! Li0 reaction in fair agreement with the
work of Chamelot et al.93

Figure 20 Standard potential in LiF–BeF2 (66–34) relative
to the HF(g)/H2 couple calculated at T¼1000 K.


values as a function of temperature as given in
Table 6, which gives the standard potentials for the
main salt carrier elements, the actinides, and some
elements of structural materials. Figure 20 shows the
electrochemical potentials calculated for T ¼ 1000 K.
In a recent study, Chamelot et al.93 studied the
electrochemical potentials of ThF4, NdF3, and
GdF3 in the LiF–CaF2 (77–23) solvent in order to
demonstrate the reprocessing scheme of the molten
salt fuel. The LiF–CaF2 system has been selected in
their study as it has a lower melting point compared
to pure LiF. The experimental results are given in
Table 7 and show that the LiF–CaF2 (77–23) solvent
can be alternatively used to reduce Th, Nd,
and Gd from this salt as the redox potentials of
Mxþ þ xeÀ ! M0 (M ¼ Th, Nd, Gd) reactions are
more positive than in the case of the Liþ þ eÀ ! Li0
reaction and so are reduced prior to the solvent.
These authors also concluded that the LiF–BeF2
(67–33) composition, as the typical MSR carrying

3.13.7 Radiation Stability of
Molten Salts
As in ceramic fuels, the fuel carrier in a MSR will be
subjected to various types of radiation that can cause
damage, such as a- and b-decay, g-radiation, and
neutron and fission products. But unlike ceramic
fuels, a liquid does not have a lattice structure
(long-range order) that can be distorted.
As reported by Blankenship,95 radiolytic formation of F2 occurs in the fluoride salts at low temperatures (T < 100  C), but, because all the salts

considered as MSR fuel are in the solid state at
these temperatures, the evolution rate is somehow
limited by a slow fluorine diffusion within the crystal.
At higher temperatures, a reverse reaction counteracts primary radiolysis events, which happens for
most of the salts far below their melting points.
It has been demonstrated that, during this recombination process, F2 reacts more rapidly with salts that
have primarily lost their fluorine atoms and, thus, the


Molten Salt Reactor Fuel and Coolant

F2 buildup in the reactor is eliminated.95 Because the
MSR operates at high temperatures, the recovery
process is rapid and radiation damage to the salt is
very small. This has been confirmed in separate
experiments, using accelerators, and in in-pile tests
for the ARE and MSRE projects. None of these experiments have revealed indications that the fluoride salts
are unstable in radiation fields.8,95 It is believed that
this radiation stability is responsible for the demand
that only very stable salts must be considered in the
reactor in order to keep the construction alloys thermodynamically stable with respect to the salt.

3.13.8 Fission Product Behavior
The fission products that are formed during the
operation of the MSR can be divided into three
main groups based on their solubilities in the carrying matrix: noble gases, stable salt-soluble fluorides,
and noble metals that are very difficult to dissolve in
the fluoride matrix. Whether the fission product will
or will not be dissolved by the salt is determined by
the redox potential of the salt. As demonstrated in

the MSRE project, the redox potential of the salt is
controlled by the UF4/UF3 ratio in such way that
the corrosion of the structural material, for example,
leaching of chromium (the least stable element

383

in the Ni-based alloys, see Section 3.13.9) from
the Hastelloy-N,83 is inhibited. As reported by
Rosenthal et al.,80 the UF4/UF3 ratio in the MSRE
was $100. It is shown in Figure 21 that at this
concentration the ratio of dissolved chromium in
the form of CrF2 and its metal form is <10À5.
Taking into account that the UF4/UF3 ratio is set
in such way not to form chromium fluoride, one
can assume that fluorides that have more negative
free energy of formation DGf0 according to the
general reaction:
xMmetal þ yF2gas ! xMF2y

DGf0

½IIŠ

will dissolve in the fuel, whereas the ones that have
higher DGf0 of the above given reaction will precipitate in the form of insoluble metals.
During the operation of the MSR, free fluorine is
formed from the fission processes. This fluorine preferably reacts with UF3, increasing the UF4/UF3 ratio
and thus changing the redox potential of the
salt. This will certainly increase the corrosion rate

of the structural material; therefore, the UF3 concentration must be readjusted. This is achieved by adding
small amounts of pure metals, for example, beryllium,
which absorb fluorine. In the MSRE, a beryllium
rod was kept immersed in the salt until the UF3
concentration reached the correct value. On the
other hand, as discussed in Section 3.13.9, too high

100
10–1

/F
e

/N
i
Ni

F

2

Fe

2

/C

r

F


2

10–3
F

10–4

Cr

10–5
Mo

MFx / M mole ratio

10–2

Mo
F3 /

10–6
10–7
10–8
100

101

102

103

104
105
UF4 / UF3 mole ratio

106

107

Figure 21 Variation of equilibrium concentration of structural metal fluorides as a function of the UF4/UF3 ratio in a molten
salt reactor fuel. Reproduced from Rosenthal, M. W.; Haubenreich, P. N.; Briggs, R. B. Tech. Rep. ORNL-4812; 1972.